In this paper, the Da Lat Nuclear Research Reactor (DNRR) core has been modeled by SCALE/TRITON code to generate two-group homogenized cross-sections for 3D kinetics calculations. In the calculation, plate-type model has been applied in selfshielding structure while the fuel assemblies have been grouped for the cross-section generation.
PRELIMINARY STUDY ON MODELING THE DALAT NUCLEAR RESEARCH REACTOR AND GENERATING THE MULTI-GROUP CROSS-SECTION FOR THREE DIMENSIONAL REACTOR KINETICS CALCULATIONS Ta Duy Long1,*, Nguyen Hoang Tu2, Nguyen Thi Dung Institute for Nuclear Science and Technology, 179 Hoang Quoc Viet, Ha Noi Vietnam Agency for Radiation and Nuclear Safety, 113 Tran Duy Hung, Ha Noi *Email: duylong09@gmail.com Abstract: Nowadays nuclear kinetic codes coupled with thermal-hydraulics codes are necessary in analyzing transients/accidents scenarios to ensure the safety of the reactor The 3D kinetic code PARCS is used for the DNRR using homogenized macroscopic cross-section to be prepared by other physics lattice codes such as MCNP, SCALE, SERPENT In this paper, the Da Lat Nuclear Research Reactor (DNRR) core has been modeled by SCALE/TRITON code to generate two-group homogenized cross-sections for 3D kinetics calculations In the calculation, plate-type model has been applied in selfshielding structure while the fuel assemblies have been grouped for the cross-section generation The preliminary calculation results not only include the cross-sections for incore region but also for the reflective graphite region for kinetic calculations Keywords: SCALE, TRITON, homogenized cross-section, DNRR NGHIÊN CỨU SƠ BỘ VỀ MÔ HÌNH HĨA VÙNG HOẠT LỊ ĐÀ LẠT VÀ PHÁT DỮ LIỆU TIẾT DIỆN BẰNG CHƢƠNG TRÌNH SCALE/TRITON SỬ DỤNG CHO TÍNH TỐN ĐỘNG HỌC LỊ PHẢN ỨNG BA CHIỀU Tóm tắt: Ngày việc sử dụng chƣơng trình tính tốn động học kết hợp với chƣơng trình tính toán thủy nhiệt trở nên cần thiết tốn phân tích trạng thái chuyển tiếp nhằm đảm bảo an tồn vùng hoạt lị phản ứng Chƣơng trình tính tốn động học lị phản ứng PARCS sử dụng liệu tiết diện vĩ mô đồng đƣợc tính tốn chƣơng trình tính tốn lƣới mạng nhƣ MCNP, SCALE, SERPENT Trong báo cáo này, lị phản ứng hạt nhân Đà Lạt đƣợc mơ hình hóa bời chƣơng trình SCALE/TRITON nhằm đƣa tiết diện hai nhóm sử dụng cho tính tốn động học Trong tính tốn này, mơ hình phẳng đƣợc sử dụng tính tốn tự che chắn bó nhiên liệu, bó nhiên liệu đƣợc phân loại theo đặc trƣng cho liệu tiết diện đồng đƣợc tính tốn Các kết sơ báo cáo đƣợc tính tốn cho bên ngồi vùng hoạt đến lớp vành phản xạ graphit lò phản ứng Đà Lạt Từ khóa: SCALE, TRITON, homogenized cross-section, DNRR 1 INTRODUCTION Nowadays, the coupled calculations between neutronics and thermal-hydraulics codes become popular in nuclear reactor calculations to consider the thermal-hydraulics feedbacks [1-2] Along with the development of the computer and the computational tools, the kinetic codes are also taking into account in coupling calculation with the system codes to analyze transient or accident conditions as well as incidents [3-5] which can occur in a nuclear reactor Among the nuclear reactor kinetics calculation codes, the PARCS code which was developed by Purdue University, USA and used by USNRC in analyzing the transient states in the reactor, based on nodal method and diffusion theory for the threedimensional multi-group kinetics calculation is one of the most popular tools in nuclear reactor safety analysis The PARCS code, which has the ability to couple with thermal hydraulics code like TRACE, RELAP5 are widely used in analyzing the transients and accidents in both the nuclear power reactors [3-5] and nuclear research reactors [6-7] In PARCS code, the homogenized macroscopic cross-sections for the kinetic calculations are provided by other lattice codes such as SCALE/TRITON, SERPENT or MCNP, etc The TRITON code in the SCALE code system, which is a powerful tool for lattice physics calculations [8] is normally used in generating cross-section for PARCS by its accuracy in lattice physics calculation and the capability to couple with PARCS [9] The DNRR, which based on TRIGA Mark II reactor, is the unique reactor of Vietnam until present So far, kinetic calculations for the DNRR have been performed using RELAP5 code with point kinetic model [10-12], while the 3D kinetic calculations for the DNRR to compare with the experimental results have been still under studying This research on homogenized cross-section generating will be used as the first step for the 3D kinetic model calculation using PARCS for analyzing both the steady and the transient conditions for the DNRR In this paper, the DNRR fuel assemblies and reflector structures are modeled by SCALE/TRITON The fuel assemblies are grouped based on their positions to acquire the average cross-section of each group, the structure materials are also grouped according to their material compositions Due to the limitation of selfshielding structure in TRITON for the DNRR fuel assembly geometry, a plate-type model [7] which has been applied for a research on VR-1 reactor will be used Then, the homogenized macroscopic cross-sections of the fuel assemblies and other structure cells like beryllium at neutron trap, auxiliary beryllium block reflector around the reactor core and graphite reflector are generated THE DALAT NUCLEAR RESEARCH REACTOR MODEL IN SCALE/TRITON In this research, the Low Enriched Uranium (LEU) core of the DNRR is considered The fuel assembly of the DNRR is VVR-M2 fuel type, consist UO2-Al alloy fuel elements with 0.94 mm of fuel meat thickness and two cladding layers with the thickness of one layer 0.78 mm, as shown in Fig Fig 1: The DNRR fuel assembly The two inner fuel elements of the DNRR fuel assembly are cylinder shape while the outer fuel element has the hexagonal geometry The active length of the fuel assembly is 600 mm The material specification of the UO2-Al fuel using in DNRR core is shown in Table Table 1: Material specification of LEU fuel of the DNRR Nuclide 234 U Density (Atom/barn/cm) 1.34219E-05 235 U 1.19978E-03 238 U 4.80027E-03 16 O Al 1.20269E-02 4.16117E-02 The DNRR core has diameter of 44.2 cm and consists of 92 fuel assemblies with 19.75 wt% of 235U The neutron trap is located at the center of the reactor core, surrounded with layers of Beryllium block, each block has the same size as the fuel assembly There are irradiation channels (1-4, 7-1 and 13-2), shim rods, safety rods and an automatic rod in the reactor core The reflective graphite region around the reactor core has the inner and outer radius are 23.75 cm and 54.25 cm, respectively The rotary specimen consists of 40 irradiation holes with the diameter of 31.75 cm, located at the reflective graphite region and it is also taken into account in the calculation model For 3D kinetic calculation with PARCS code, a 3D homogenized cross-section must be prepared by using SCALE/TRITON Due to the limitation of 2D model in TRITON, the 3D cross-section is prepared by using a sequence of 2D models at different axial layers of the DNRR In this paper, a 2D model of DNRR is presented at the axial layer near the center of the reactor core, in which the rotary specimen is located at the graphite reflector region The 2D DNRR core geometry and the graphite reflector region are shown in Fig 2, the details are given in the Safety Analysis Report (SAR) of the Da Lat Nuclear Research Reactor [13] Fig 2: The DNRR core and reflective structure Due to the limitation of the self-shielding structure with the geometry similar to the DNRR fuel assembly, the plate-type model with the conservation of the fuel and cladding thickness was applied when the half-pitch (HPITCH) was interpolated based on the results of SRAC2006 calculation for the actual model of the fuel assembly [7] The infinite multiplication factor (k-inf) for different half-pitches of plate-type model are calculated by SRAC2006 and compared with the result of the 2D actual fuel assembly model to determine the half-pitch which can be used in self-shielding specification of SCALE/TRITON The results of the infinite multiplication factor versus the change on the half-pitch of the plate-type model are shown in Table Comparing with the k-inf value of actual model, k-inf = 1.63557, the half-pitch value for the self-shielding structure was chosen with the value of HPITCH=0.592 cm Table 2: k-inf versus half-pitch for the plate-type model calculated by SRAC HPITCH k-inf 0.59 0.591 0.592 0.593 0.594 0.595 0.596 0.597 0.598 0.599 1.63618 1.63594 1.63569 1.63543 1.6352 1.63494 1.63471 1.63446 1.63422 1.63397 1.6365 1.636 1.6355 1.635 1.6345 1.634 1.6335 1.633 1.6325 0.59 0.591 0.592 0.593 0.594 0.595 0.596 0.597 0.598 0.599 Fig 3: k-inf versus half-pitch for the plate-type model calculated by SRAC In kinetic calculation within PARCS code, the homogenized macroscopic crosssections of fuel assemblies to be prepared by TRITON code are used as the input parameters To simplify the model in neutron kinetic calculations, 92 fuel assemblies in the DNRR reactor core were divided into groups by their different positions in the reactor core [14] In details, fuel assembly groups are surrounded by: - Group 1: Surrounded by shim rods or safety rods, beryllium block and other fuel assemblies Group 2: Surrounded by beryllium block and other fuel assemblies Group 3: Surrounded by shim rods or safety rods, irradiation channel and other fuel assemblies Group 4: Surrounded by shim rods or safety rods and other fuel assemblies Group 5: Surrounded by automatic rod (AR) and other fuel assemblies Group 6: Surrounded by irradiation channel and other fuel assemblies - Group 7: Surrounded by only other fuel assemblies For preparing input file of PARCS code, the homogenized macroscopic cross-sections of other structures as non-fuel are also taken into account in SCALE/TRITON code including: Neutron trap, irradiation channels, shim rods and safety rods, automatic rod, beryllium blocks around the reactor core and graphite reflector These homogenized cross-sections are calculated for both fast and thermal neutron energy groups PRELIMINARY CALCULATION RESULTS As mentioned in the previous section, the DNRR has been modeled by using TRITON code in the SCALE code system The extending of DNRR calculation model to the graphite reflector region is shown in Fig 4a while the details of the reactor core are shown in Fig 4b Fig 4: The DNRR model in SCALE/TRITON (a) and the core in details (b) In calculation model, all the shim rods and safety rods are fully inserted in the reactor core and the effective multiplication factor (k-eff) calculated by SCALE/TRITON has value of k-eff = 0.95780 The fast and thermal neutron flux of the DNRR with LEU core are also calculated by SCALE/TRITON and shown in Fig In Fig 5a, it can be seen that the fast neutron flux has the higher value in the fuel regions around the center beryllium trap, at the positions where the shim rods and safety rods are not presented Because of slowing down ability of the beryllium in the center beryllium trap, the fast neutron flux inside the center hole of the reactor are significant reduced compared to the fast neutron flux in the outer beryllium layer of the center trap The fast neutron flux outside the reactor core is reduced by the beryllium blocks located around the fuel assembly in the reactor core The thermal neutron flux distribution is shown in Fig 5b The highest value of the thermal neutron flux is located at the neutron trap of the core, where water hole is located and surrounded by beryllium blocks and beryllium rods The thermal flux is significant higher than the average at the position of the irradiation channels and has the lowest value at the position of the shim rods and safety rods Both the fast and thermal neutron fluxes have rather low value outside the reactor core in radial direction Fig 5: Fast neutron flux (a) and thermal neutron flux (b) distribution in the DNRR Table presents the homogenized macroscopic cross-sections by using the SCALE/TRITON calculation These cross-sections include transport, absorption, scattering cross-sections of fast and thermal energy as well as the fission cross-sections within two energy groups For the kinetic calculation with PARCS code, the crosssections are calculated for fuel assemblies and also the other structure regions inside the reactor core However, in the paper, only one axial layer of the DNRR has been calculated, the remaining axial layers for generating homogenized cross-sections for 3D kinetic calculations by using TRITON are under carried out Table 3: The homogenized macroscopic cross-section generated by SCALE/TRITON for the DNRR Homogenized macroscopic cross-section Fast neutron energy group Cell regions Fuel group Fuel group Fuel group Fuel group Fuel group Fuel group Fuel group Automatic rod Irradiation channel Outer neutron trap layer Inner neutron trap layer Shim and safety rods Beryllium block and reactor core tank Graphite Reflector region Thermal neutron energy groups downscattering 2.29E-02 Transport Absorption nu-fission 1.03E+00 9.22E-02 Fission Transport Absorption nu-fission 2.07E-01 5.00E-03 4.13E-03 kappafission 5.20E-14 Fast Thermal 1.64E-01 kappafission 2.10E-12 1.67E-03 6.75E-02 2.07E-01 5.11E-03 4.22E-03 5.32E-14 2.39E-02 1.01E+00 8.99E-02 1.60E-01 2.05E-12 1.71E-03 6.58E-02 2.04E-01 5.04E-03 4.18E-03 5.26E-14 2.37E-02 1.01E+00 9.03E-02 1.61E-01 2.05E-12 1.69E-03 6.61E-02 2.04E-01 4.91E-03 4.10E-03 5.16E-14 2.27E-02 9.86E-01 8.76E-02 1.56E-01 1.99E-12 1.66E-03 6.40E-02 2.07E-01 5.17E-03 4.27E-03 5.38E-14 2.45E-02 9.91E-01 8.85E-02 1.58E-01 2.01E-12 1.73E-03 6.48E-02 2.05E-01 5.12E-03 4.23E-03 5.33E-14 2.42E-02 1.02E+00 9.16E-02 1.64E-01 2.09E-12 1.71E-03 6.71E-02 2.03E-01 4.94E-03 4.14E-03 5.21E-14 2.34E-02 9.90E-01 8.79E-02 1.57E-01 2.00E-12 1.67E-03 6.43E-02 2.73E-01 2.09E-03 0.00E+00 0.00E+00 1.80E-02 1.17E+00 7.83E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.31E-01 3.50E-04 0.00E+00 0.00E+00 4.90E-02 1.62E+00 1.51E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.89E-01 1.18E-03 0.00E+00 0.00E+00 6.48E-03 7.62E-01 1.35E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.67E-01 7.60E-04 0.00E+00 0.00E+00 3.24E-02 1.39E+00 9.39E-03 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.41E-01 5.41E-02 0.00E+00 0.00E+00 1.38E-02 1.69E+00 4.19E-01 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.88E-01 1.27E-03 0.00E+00 0.00E+00 5.41E-03 6.47E-01 5.23E-02 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.56E-01 2.78E-05 0.00E+00 0.00E+00 2.69E-03 3.82E-01 2.40E-04 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CONCLUDING REMARKS In the paper, the calculation model and results of the DNRR for generating homogenized macroscopic cross-sections using SCALE/TRITON have been presented For preparing cross-section data to the kinetic calculations, fuel assemblies inside reactor core of the DNRR are divided into groups based on their position specifications The other structure region cross-sections are also calculated for using in the kinetic code Because of the limitation of self-shielding specification in SCALE with the geometry of the DNRR fuel assembly, plate-type model for self-shielding structure in SCALE/TRITON is used for analyzing the VR-1 reactor, is also applied for the DNRR fuel assembly in cross sections calculation The calculation results of SCALE/TRITON code include multiplication factor, neutron flux distribution and homogenized cross-sections that are calculated within two neutron energy group for transport, absorption, scattering and fission cross sections These cross sections will be used for the kinetic calculations with PARCS code However, for using in the 3D kinetic calculation, the homogenized cross-sections are need to be calculated for all axial layers of the DNRR and also need to be verified by experiment of calculation results from other code as MCNP5 REFERENCES [1] Coupled fine-mesh neutronics and thermal hydraulics – Modeling and implementation for PWR fuel assemblies, KlasJareteg et.al, 2015 [2] Neutronics and sub-channel thermal-hydraulics analysis of the Iranian VVER-1000 fuel bundle, F Faghihi et.al, 2015 [3] Analysis of the OECD MSLB Benchmark with the Coupled Neutronic and ThermalHydraulics Code RELAP5/PARCS, Kozlowski T et.al, 2000 [4] Simulation of rod ejection accident in a WWER-1000 Nuclear Reactor by using PARCS code, Noori-Kalkhoran O et.al, 2014 [5] Full Scope Thermal-Neutronic Analysis of LOFA in a WWER-1000 Reactor Core by Coupling PARCS v2.7 and COBRA-EN, Noori-Kalkhoran O et.al, 2014 [6] Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor, Siefman D J et.al, 2015 [7] Deterministic and Stochastic Neutron Cross-section Generation for PARCS Applied to VR-1 Reactor, Fejt F., 2016 [8] High-fidelity lattice physics capabilities of the SCALE code system using TRITON, Mark D DeHart et.al, 2007 [9] Lattice physics capabilities of the SCALE code system using TRITON, Mark D DeHart, 2006 [10] Some results obtained from the use of reactivity meter to measure the integral characteristics of the control rods on the Dalat Nuclear Research Reactor, Nguyen NhiDien et.al, 1997 [11] Kinetics characteristics of the Dalat Nuclear Research Reactor, Tran Khac An et.al, 1993 [12] Application of RELAP5/MOD3.2 Code for the Dalat Nuclear Research Reactor, Le VinhVinh, Huynh Ton Nghiem, 2006 [13] Báo cáo phân tích an tồn lị phản ứng Hạt nhân Đà Lạt, 2012 [14] Macroscopic Cross Section Generation with SCALE 6.2 for the MYRRHA Minimal Critical Core - International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering, Korea - 2017 [15] SCALE code system manual, 2017 [16] Nguyen KienCuong, Da Lat Nuclear Research Reactor, Private Message, 2018 ... analyzing the transient states in the reactor, based on nodal method and diffusion theory for the threedimensional multi-group kinetics calculation is one of the most popular tools in nuclear reactor. .. research on homogenized cross-section generating will be used as the first step for the 3D kinetic model calculation using PARCS for analyzing both the steady and the transient conditions for the. .. INTRODUCTION Nowadays, the coupled calculations between neutronics and thermal-hydraulics codes become popular in nuclear reactor calculations to consider the thermal-hydraulics feedbacks [1-2] Along